Nuclear reactors generally use zirconium-based alloys for structural components of fuel assemblies in the reactor core, e.g. fuel cladding, end plugs, flow channels, and spacers. Zirconium and its alloys, under normal circumstances, are excellent structural material for these components because they have low neutron absorption cross sections and, at temperatures below about 350.degree. C., are strong, ductile, extremely stable and relatively nonreactive in the presence of demineralized water or steam. "Zircaloys" are a family of corrosion-resistant zirconium alloy cladding materials. They are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used alloys for cladding. Zircaloy-2 has on a weight basis about 1.2 to 1.7 percent tin; 0.13-0.20 percent iron; 0.06-0.15 percent chromium and 0.05 to 0.08 percent nickel. Zircaloy-4 has essentially no nickel and about 0.2% iron but is otherwise substantially similar to Zircaloy-2.
During normal service, these materials can pick up hydrogen resulting from the corrosion of the zirconium alloys in water or steam environments. As the hydrogen concentration of the material increases, its ductility decreases and thereby establishes conditions for potential damage to the fuel components. This effect is most pronounced at moderate temperatures, e.g. &lt;200.degree. C., as experienced by the fuel during reactor shutdown and fuel handling operations.
A particularly deleterious form of hydride embrittlement occurs in the interior of a fuel cladding when water enters through a breach in the cladding wall. The initial "primary defect" through which the water enters is not necessarily catastrophic, and the breached cladding can sometimes be used for a period of time. However, the damaged cladding is susceptible to localized hydriding and ensuing "secondary defects". Secondary defects can be much worse than their primary counterparts, often allowing release of large amounts of fission products and fuel material from erosion/corrosion of the fuel material.
Secondary defects likely form according to the following sequence of events. When the fuel rod is initially breached, the coolant water enters the tube and instantly flashes to steam. Some of the steam probably reacts with the zirconium on the inner surface of the cladding to give off hydrogen, some which is gaseous and some of which is absorbed by the zirconium. If unchecked, both the absorbed and gaseous hydrogen can lead to formation of zirconium hydride. However, the gaseous hydrogen is apparently directly responsible for the potentially catastrophic localized hydriding. When the hydrogen gas reaches a critical concentration at which it significantly displaces oxygen in the atmosphere (a condition sometimes referred to as oxygen starvation), large amounts of zirconium hydride will form locally on a bare zirconium (or zirconium alloy) surface, or a break in the thin oxide layer on the zirconium (or zirconium alloy) often appearing as a "blister" or "sunburst."
The directly absorbed hydrogen also damages zirconium components. However, it does so by forming brittle zirconium hydride platelets within the base metal matrix. The platelets generally form first on grain boundaries, and continue to grow with increasing hydrogen content until a continuous hydride phase at the grain boundaries is achieved. They preferentially form on the basal planes of alpha phase processed metals.
Various approaches have been taken to prevent the formation of zirconium hydride plates in structural elements for nuclear fuel applications. For example, yttrium and some other rare earth metals, when in intimate contact with zirconium, have been found to act as preferential sinks for hydrogen. A fuel rod design incorporating monolithic layers of yttrium provided as inner liners on a zirconium-based alloy tube, or as an internal cap on an end plug is described in German Auslegeschrift 19 03 009, published on Dec. 16, 1976. Unfortunately, any bare yttrium used in such tubes would likely corrode in the fuel core environment, thus degrading its usefulness in preventing hydride formation.
U.S. Pat. No. 4,659,545 (issued to Ferrari on Apr. 21, 1987) describes a similar design in which between 1-40 percent of a cladding inner surface area is coated with thin film nickel spots. This design supposedly provides multiple hydrogen entry regions (the nickel film spots) so as to distribute the hydrogen more uniformly over the inner surface of the cladding. Unfortunately, this design still allows formation of locally high hydrogen concentrations in the zirconium alloy matrix near the nickel spots. Thus, hydride platelets can still form in the cladding tube, albeit at separated regions.
U.S. Pat. 5,475,723 (Application No. 08/215,451, Attorney Docket No. 24-NT-05494/GENEP011, entitled "NUCLEAR FUEL CLADDING WITH HYDROGEN ABSORBING LINER", naming Marlowe as inventor, filed on Mar. 16, 1994, and assigned to the assignee hereof) describes an improved fuel cladding design having a hydrogen-absorbing zirconium-alloy inner liner. The zirconium alloy contains a matrix having a relatively high concentration of a hydrogen absorbing material such as nickel. Thus, hydrogen in a fuel rod is absorbed uniformly over the cladding interior. Although this design represents a significant advance over the prior art, it would still be desirable to have other designs in which hydrogen is prevented from reacting with the zirconium alloy matrix.